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In state-of-the-art stellarators, turbulence is a major cause of the degradation of plasma confinement. To maximize confinement, which eventually determines the amount of nuclear fusion reactions, turbulent transport needs to be reduced. Here we report the observation of a confinement regime in a stellarator plasma that is characterized by increased confinement and reduced turbulent fluctuations. The transition to this regime is driven by the injection of submillimetric boron powder grains into the plasma. With the line-averaged electron density being kept constant, we observe a substantial increase of stored energy and electron and ion temperatures. At the same time, the amplitude of the plasma turbulent fluctuations is halved. While lower frequency fluctuations are damped, higher frequency modes in the range between 100 and 200 kHz are excited. We have observed this regime for different heating schemes, namely with both electron and ion cyclotron resonant radio frequencies and neutral beams, for both directions of the magnetic field and both hydrogen and deuterium plasmas.

Recently an improved confinement regime, characterized by reduced turbulent fluctuations has been observed in the Large Helical Device upon the injection of boron powder into the plasma (Nespoli et al 2022 Nat. Phys. 18 350–56). In this article, we report in more detail the experimental observations of increased plasma temperature and the decrease of turbulent fluctuations across the plasma cross section, on an extended database. In particular, we compare powders of different materials (B, C, BN), finding similar temperature improvement and turbulence response for the three cases. Modeling of the powder penetration into the plasma and of neoclassical electric field and fluxes support the interpretation of the experimental results. Additionally, we report evidence of the temperature improvement increasing with powder injection rates and decreasing for both increasing density and heating power. Though, plasma turbulence response varies depending on the initial conditions of the plasma, making it difficult to draw an inclusive description of the phenomenon.

The cracking thresholds were evaluated for tungsten samples with different microstructure in the course of QSPA Kh-50 repetitive plasma loads. No damage has been observed on the exposed surfaces under 0.1 MJ/m2. Nevertheless, cracks were detected in the bulk of irradiated tungsten (with longitudinal grain orientation). Increasing heat load up to 0.2 MJ/m2 caused the damaging of all types of tungsten targets. The observed cracks propagate to the bulk mainly transversely and parallel to the irradiated surface. The effect of the subsequent exposure with LHD divertor plasma on the tungsten samples was analyzed. The obtained results are discussed.

In order to explore the compatibility of good core plasma performance with divertor heat load mitigation, the interaction between cold edge plasma and core plasma transport, including the edge transport barrier (ETB), has been analysed in the divertor detachment discharges of deuterium plasmas in LHD with resonant magnetic perturbation (RMP) field application. The RMP application introduces a widened edge stochastic layer and sharp boundary in the magnetic field structure between the confinement region and the edge stochastic layer. The widened edge stochastic layer enhances impurity radiation and provides stable detachment operation as compared with the case without RMP. It is found that ETB is formed at the confinement boundary at the onset of detachment transition. However, as the detachment deepens, the resistive pressure gradient-driven MHD mode is excited, which degrades the ETB. At the same time, however, the core transport decreases to keep global plasma stored energy (W p) unchanged, showing clear core-edge coupling. After a gradual increase of density fluctuation during the MHD activity, a spontaneous increase of W p and the recovery of ETB are observed while the detachment is maintained. Then, the coherent MHD mode ceases and ELM-like bursts appear. In the improved mode, impurity decontamination occurs, and the divertor heat load increases slightly. Key controlling physical processes in the interplay between core and cold edge plasma are discussed. A comparison between deuterium and hydrogen plasmas shows that hydrogen plasmas exhibit similar features to the deuterium ones in terms of density and magnetic fluctuations, impurity decontamination towards higher confinement, etc. But most of the features are modest in the hydrogen plasmas and thus no clear confinement mode transition with clear ETB formation is defined. Better global confinement is obtained in the deuterium plasmas than the hydrogen ones at a higher radiation level.

Tritium (T) retention characteristics in dust collected from the divertor in JET with ITER-like wall (JET-ILW) after the third campaign in 2015–2016 (ILW-3) have been examined in individual dust particles by combining radiography (tritium imaging plate technique) and electron probe micro-analysis. The results are summarized and compared with the data obtained after the first campaign in 2011–2012 (ILW-1). The dominant component in ILW-1 dust was carbon (C) originating from tungsten-coated carbon fibre composite (CFC) tiles in JET-ILW divertor and/or legacy of C dust after the JET operation with carbon wall. Around 85% of the total tritium retention in ILW-1 dust was attributed to the C dust. The retention in tungsten (W) and beryllium (Be) dominated particles was 100 times smaller than the highest T retention in carbon-based particles. After ILW-3 the main component contributing to the T retention was W. The number of small W particles with T increased, in comparison to ILW-1, most probably by the exfoliation and fragmentation of W coatings on CFC tiles though T retention in individual W particles was smaller than in C particles. The detection of only very few Be-dominated dust particles found after ILW-1 and ILW-3 could imply stable Be deposits on the divertor tiles.

Heat transfer through forced convection from the inner surface of a circular tube to force the flow of liquid sodium in the laminar and transition regions were numerically analysed for two types of tube geometries (concentric annular and circular tubes) and two types of equivalent diameters (hydraulic and thermal equivalent diameters). The unsteady laminar three-dimensional basic equations for forced convection heat transfer caused by a step heat flux were numerically solved until a steady state is attained. The code of the parabolic hyperbolic or elliptic numerical integration code series (PHOENICS) was used for calculations by considering relevant temperature dependent thermo-physical properties. The concentric annular tube has a test tube with inner and outer diameters of 7.6 and 14.3 mm, respectively, has a heated length of 52 mm, and an L/d of 6.84. The two circular tubes have inner diameters of 6.7 and 19.3 mm with L/d of 7.76 and 2.69, respectively, and a heated length of 52 mm. The inlet liquid temperature, inlet liquid velocity, and surface heat flux were equally set for each test tube as T(in)≅573 to 585 K, u(in) = 0.0852 to 1 m/s, and q = 2x10(5) to 2.5x10(6) W/m(2), respectively. The increase in temperature from the leading edge of the heated section to the outlet of the circular tubes (with a hydraulic diameter of d(H) = 6.7 mm and a thermal equivalent diameter d(te) = 19.3 mm) was approximately 2.70 and 1.21 times as large as the corresponding values of the concentric annular tube with an inner diameter of 7.6 mm and an outer diameter of 14.3 mm, respectively. A quantity in the laminar and transition regions was suggested as the dominant variable involved in the forced convection heat transfer in the circular tube. The values of the local and average Nusselt numbers, Nu(z) and Nu(av), respectively, for a concentric annular tube with d(H) = 6.7 mm and for a circular tube with d(H) = 6.7 mm were calculated to examine the effects of q, T(in), and Pe on heat transfer. The general correlation in the laminar and turbulent regions for a high temperature liquid metal was identified for the circular tube with a hydraulic diameter.

The advanced fabrication method named advanced multi-step brazing (AMSB) has been developed for fabrication of a tungsten (W)/copper alloy divertor heat removal component in a fusion reactor. The principle of AMSB is repetitive application of the advanced brazing technique (ABT). The ABT was originally developed in our previous work for jointing of W and oxide dispersion strengthened copper alloy (ODS-Cu: GlidCop®) with the BNi-6 (Ni-11%P) filler material. The special feature of the AMSB is that the leak-tight joint of GlidCop® (GlidCop®/GlidCop®) and of stainless steel (SUS) and GlidCop® (SUS/GlidCop®) can be realized by application of the ABT. Therefore, the AMSB enables production of a leak-tight sealing structure with the appropriate lid material such as GlidCop® or SUS for a pre-processed rectangle-shaped cooling flow path channel. The rectangle-shaped cooling flow path channel has an advantage regarding its heat removal capability. Another special feature of the AMSB is that the physical properties of the AMSB joint has strong tolerance against the repetitive brazing heat cycle. Thus, repetitive application of the ABT does not cause any undesired effects on the leak tightness of the GlidCop®/GlidCop® and the SUS/GlidCop®. The new AMSB type component with the rectangle-shaped fluid flow path and the V-shaped staggered rib structure were successfully produced; therein a pre-processed rectangle-shaped cooling flow path channel was sealed with a GlidCop® and SUS lid structure with leak-tight conditions. The component showed excellent heat removal capability under reactor-relevant conditions with ∼30 MW m−2.

A core density collapse (CDC) phenomenon is a rapid collapse events observed in super dense core (SDC) plasma with internal diffusion barrier (IDB) in the Large Helical Device (LHD). By CDC, the central beta is decreased by up to 50%. The collapse starts from the edge region of the plasma. CDCs appear with plasma parameters where the high–n ballooning modes are unstable at ϱ ∼ 0.8. With less collisional conditions, m = 1 type oscillations are observed with similar beta profile. The origin of the m = 1 oscillations is not clarified (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

To understand the edge plasma transport in the complex magnetic field lines structure in the LHD heliotoron, the heat and particle fluxes profiles on divertor plates were investigated experimentally and theoretically by using EMC3-EIRENE code. The heat flux profiles on a divertor plate in different magnetic configuration were measured and the main channel of the heat transport was revealed to be the field lines with long connection lengths which connect the divertor plate and the stochastic field lines layer. A comparison of the heat and the ion fluxes profiles on the divertor plates between the experimental data and the results of the calculation using EMC3-EIRENE code was conducted. It was revealed that the change of the cross-field transport in the edge plasma region causes not only the change of the heat and particle flux profiles on divertor plates, but also the changes of the intrinsically non-uniform global heat and particle deposition profiles on the helical diveror. The cross-field transport seems to be larger in the higher edge electron temperature discharge in the experimental conditions in this paper (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

First walls and divertor plates of large helical device (LHD) are made of stainless steels and isotropic graphite, respectively. They are bombarded by energetic plasma particles, and are caused the radiation damages. For understanding of long term irradiation effects of metallic first walls, probe specimens of Mo were mounted on the special sample stage of outer-port of the LHD torus which is located at about 2500 mm from LCFS (last closed flux surface) of heliotron plasma, and then, exposed to plasma discharges through the single experimental campaign. Majority of discharge gas was hydrogen or helium. Amorphous re-deposition layer of about few nm thick and large sized blisters with few micrometer were formed at the specimen surface. Formation of dense gas bubbles (1–2 nm), dislocation loops and surface roughening were simultaneously occurred in the sub-surface layer about 30 nm thick. Such nano scale damages may be mainly caused by helium particle bombardment.

The impurity transport properties in the ergodic layer of LHD are analyzed by the 3D edge transport code as well as 1D simple model, which is used to illustrate the essential transport terms in the analysis. It is found that as the plasma density increases the edge surface layers, very edge region of the ergodic layer, enters friction dominant regime, resulting in impurity retention. It is considered that the cause for the retention is both temperature drop and the flow acceleration in the edge surface layers. The edge surface layers can provide effective retention of impurities coming from divertor as well as the first wall, because of the geometrical advantage of the edge region of LHD. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

A repetitive pellet injector has been developed and pellet refuelling experiments have been launched on the Large Helical Device (LHD) in a new regime characterized by high density steady state operation. In order to suppress the disturbance to the core plasma as much as possible, we employ relatively small and slow pellets. The size and velocity of the pellets are typically 2.5?mm in diameter and 250?m?s?1, respectively. The pellets were injected into neutral beam heated plasmas with repetition frequency of 10?Hz. The density rise per pellet is 1.5 ? 1019?m?3 and the pellet penetration depth, which is normalized by the plasma minor radius, is typically less than 0.5. Quasi-steady-state operation was achieved at plasma parameters of , Te = Ti = 1.0?keV. Repetitive pellet fuelling showed improved energy confinement compared with massive gas puff fuelling in the high density regime in common with previous transient pellet injection experiments, and it can sustain such an improved confinement property.

The divertor/SOL transport of the Local Island Divertor (LID) configuration in the Large Helical Device (LHD) is studied. Since the LID configuration has an inherent three dimensional magnetic structure, the 3D edge transport code, EMC3-EIRENE, is applied in order to elucidate the 3D divertor transport physics. The main observations reported here are: 1. The effective temperature cooling due to the integrated perpendicular energy loss along the long connection length. 2. The significant plasma flow acceleration caused by the upstream source that is attributed to the geometrical significance of the LID configuration. 3. The effect of the flow acceleration on the impurity screening considering a parallel transport along the SOL.

Steady state divertor operation with high performance plasmas (ne ~ 0.7 × 1019 cm−3, Ti ~ 2 keV) was demonstrated for half an hour in the Large Helical Device (LHD), the superconducting helical device (R = 3.6–3.9 m, a = 0.6 m, B = 3 T, l/m = 2/10). The high performance plasmas have been sustained with an averaged heating power of 680 kW and achieved an injected energy of 1.3 GJ. This required both advanced technological integration of heating systems and divertor heat flux control. In particular, optimization of divertor heat flux distribution along the divertor leg trace on divertor plates and real-time magnetic axis sweeping (R = 3.67–3.7 m) have allowed LHD to access a steady state regime with a margin of safety for the actively cooled divertor plates. The distribution of divertor heat load along the traces was investigated with calorimetric measurements and it was found that there was a localized heat load connected with the loss of high-energy ions produced by ion cyclotron radio frequency near-fields. Orbit analysis shows that the behaviour of high-energy ions is qualitatively in good agreement with the experimental result. Long-pulse discharges were terminated by radiation collapse due to penetration of metallic flakes into the plasma.

Magnetic configurations of LHD are characterized by the presence of chaotic magnetic field, the so-called ergodic layer, surrounding the core plasma. H-mode-like discharges have been obtained at an outwardly shifted configuration of Rax = 4.00 m with a thick ergodic layer, where the ι/2π = 1 position is located in the middle of the ergodic layer. A clear density rise and a reduction of magnetic fluctuation were observed. ELM-like Hα bursts also appeared with a radial propagation of density bursts. These H-mode-like discharges can be triggered by changing PNBI(

On the Large Helical Device (LHD), low to high confinement (L–H) transition and edge transport barrier (ETB) formation were observed in the low beta regime (βdia 1.5%). In most of ETB plasmas electron density preferentially increases in the edge region without a substantial rise of the edge electron temperature. The ETB zone develops inside the ergodic field layer calculated in the vacuum field. The ETB formation strongly destabilizes edge coherent modes such as m/n = 2/3 or 1/2 (m, n: poloidal and toroidal mode numbers), because the plasma edge region is in the magnetic hill. The ETB is partially destroyed by the combination of these edge MHD modes and ELM-like activities. For a particular experimental condition, the forced generation of a sizable m/n = 1/1 magnetic island near the edge by application of external field perturbations facilitates the L–H transition at a lower electron density and suppresses edge MHD modes and ELM-like activities to lower levels.

Self-sustained detachment has been obtained in the Large Helical Device (LHD). Strong hydrogen gas puffing of ~200 Pa m3 s−1 after a density feedback phase detaches the plasma from the divertor plates with high reproducibility. High electron density of over 1 × 1020 m−3 is sustained without gas puffing until the heating beam stops and a high-density flat top for 2 s has been demonstrated. Throughout the self-sustained detachment phase, the minor radius of the hot plasma column shrinks to ~90% of the last-closed-flux-surface, which corresponds to the rational surface. This new state has been named the 'Serpens mode', for self-regulated plasma edge 'neath the last-closed-flux-surface. Global energy confinement of the Serpens mode is compared with the international stellarator scaling 1995 (ISS95) and the recently established scaling for high-density LHD plasmas (HD scaling), where shrinking confinement volume and shallow penetration of the heating beams are taken into account. Although the energy confinement of the Serpens mode seems deteriorated compared with ISS95, as in the case of high-density attached plasmas, it is consistent with the HD scaling. This suggests that the energy confinement properties of detached plasmas in LHD are similar to those in high-density attached plasmas.

A long-pulse plasma discharge of more than 30 min duration was achieved on the Large Helical Device (LHD). A plasma of ne = 0.8 × 1019 m−3 and Ti0 = 2.0 keV was sustained with PICH = 0.52 MW, PECH = 0.1 MW and averaged PNBI = 0.067 MW. The total injected heating energy was 1.3 GJ. One of the keys to the success of the experiment was a dispersion of the local plasma heat load to divertors, accomplished by sweeping the magnetic axis inward and outward. Causes limiting the long pulse plasma discharge are discussed. An ion impurity penetration limited further long-pulse discharge in the 8th experimental campaign (2004).

The Large Helical Device (LHD) achieves high-performance plasma confinement by the coordination of the magnetic surface region and the chaotic field line layer. It is theoretically and experimentally shown that drift surfaces exist for highly energetic particles being extended over the last closed flux surface (LCFS) in the LHD. These particles are considered lost particles due to the loss-cone in the previous theories, where the analyses are limited inside the LCFS. The present theory predicts that the loss-cone is strongly reduced in the LHD and that highly energetic particles confined over the LCFS exist. These are consistent with the LHD experimental results in both the ICRF heating experiments and the low magnetic field neutral beam injection heating experiments. From particle orbit analyses and studies on the connection length of diverter field lines, it is also shown that plasma can exist in the chaotic field line layer located outside the LCFS in the LHD. The plasma in the chaotic field line layer is clearly detected by CCD-cameras in the LHD experiment. This ambient plasma might be expected to play the role of a kind of impregnable barrier for the core plasma, which suppresses both the MHD instabilities and the cooling of the core plasma due to charge exchange processes. The line-tying effects of diverter field lines that are slipped out from the chaotic field line layer can also stabilize the ballooning mode and the vertical displacement events of the plasma column.

In the Large Helical Device (LHD), two different divertor configurations, i.e. helical divertor (HD) and local island divertor (LID), are utilized to control the edge plasma. The HD with two X-points is an intrinsic divertor for heliotron devices, accompanied with a relatively thick ergodic layer outside the confinement region. Edge and divertor plasma behavior from low density to high density regimes is presented, referring to the divertor detachment. The effect of the ergodic layer on the edge transport is also discussed. On the other hand, the LID is an advanced divertor concept which realizes a high pumping efficiency by the combination of an externally induced magnetic island and a closed pumping system. Experimental results to confirm the fundamental divertor performance of the LID are presented.

On the Large Helical Device (LHD) where nested magnetic surfaces are surrounded by the ergodic field layer, edge transport barrier (ETB) was produced in neutral-beam-injection (NBI) heated plasmas through transition and non-transition processes. The former case is the ETB formation by L-H transition, where characteristics of L-H transition observed in a tokamak plasma are clearly recognized. The confinement improvement is the modest (~ 10%), compared with the ISS95 international stellarator scaling. The threshold power for the transition is comparable or slightly lower than the ITER scaling law established by tokamaks and compact tori. The ETB is formed inside the ergodic field layer of the vacuum field. The ETB formation destabilizes edge coherent modes such as m/n = 1/1, 2/3 and 1/2, of which rational surfaces are in the magnetic hill. The formed ETB is partially and transiently destroyed by these coherent edge MHD modes and edge localized modes (ELMs) typically observed in Hα signals. The latter ETB is observed in a plasma with large reversed NBI-driven current more than 100 kA at Bt = 1 T. In these plasmas, the edge magnetic shear is enhanced by the current and the rotational transform in the core region is expected to be appreciably reduced. Thus reduced rotational transform in the plasma central region will enhance outward heat and particle fluxes toward ergodic edge layer. The ETB with steep electron temperature gradient up to ~ 5 keV/m is formed by blocking enhanced outward heat flux.

A long-pulse plasma discharge for more than 30 min. was achieved on the Large Helical Device (LHD). A plasma of ne = 0.8× 1019 m−3 and Ti0 = 2.0 keV was sustained with PICH = 0.52 MW, PECH = 0.1 MW and averaged PNBI = 0.067 MW. Total injected heating energy was 1.3 GJ, which was a quarter of the prepared RF heating energy. One of the keys to the success of the experiment was a dispersion of the local plasma heat load to divertors, accomplished by shifting the magnetic axis inward and outward.

A variety of electron density (ne) profiles have been observed in the Large Helical Device (LHD). The density profiles change dramatically with heating power and toroidal magnetic field (Bt). The particle transport coefficients, i.e. diffusion coefficient (D) and convection velocity (V) are experimentally obtained in the standard configuration from density modulation experiments. The values of D and V are estimated separately in the core and edge. The diffusion coefficients are found to be a function of electron temperature (Te), and vary with Bt. Edge diffusion coefficients are proportional to . Non-zero V is observed, and it is found that the electron temperature gradient can drive particle convection, particularly in the core region. The convection velocity both in the core and edge reverses direction from inward to outward as the Te gradient increases. However, the toroidal magnetic field also significantly affects the value and direction of V. The density fluctuation profiles are measured by a two-dimensional phase contrast interferometer. It was found that fluctuations which are localized in the edge propagate towards the ion diamagnetic direction in the laboratory frame, while the phase velocity of fluctuations around mid-radius is close to the plasma poloidal Er × Bt rotation velocity. The fluctuation level becomes larger as particle flux becomes larger in the edge region.

Glow discharge cleaning (GDC) is a widely used technique for wall conditioning in fusion experimental devices. Though the cleaning effects of GDC are essentially related to the microscopic modification of the wall surface, there are few reports about it. In the present study, specimens of wall materials were exposed to GDC plasma of helium, hydrogen and neon in the Large Helical Device (LHD) by using the retractable material probe transfer system and irradiation damage was examined by transmission electron microscopy to understand the underlying microscopic mechanism of GDC. In the case of Ne-GDC, the specimen surface was covered with a thick deposited layer of Fe and Cr but no radiation induced defects were observed. Due to the high sputtering efficiency and very shallow penetration, it is likely that neon atoms effectively sputter the surface contamination without leaving serious damage or remaining in the sub-surface region. After the Ne-GDC phase, retained Ne can be successfully removed by the following short hydrogen GDC. It was shown that a two-step GDC with Ne and H is very effective to clean the metallic surface of the LHD.

In the Large Helical Device (LHD) programme, one of the key research issues is to enhance helical plasma performance through edge plasma control. For the first time in the LHD programme, the edge plasma control was performed with a local island divertor (LID) that is a closed divertor, utilizing an m/n = 1/1 island generated externally by which 20 small perturbation coils, and fundamental LID functions were demonstrated experimentally. It was found that the outward heat and particle fluxes crossing the island separatrix flow along the field lines to the rear of the divertor head, where carbon plates are placed to receive the heat and particle loads. Accordingly, high efficient pumping was demonstrated, which is considered to be the key in realizing high temperature divertor operations, resulting in an improvement in energy confinement. In this experiment, relatively good energy confinement is achieved in the high density regime at a magnetic axis position, Rax, of 3.75?m. Results of edge modelling are also presented by using the EMC3-EIRENE code.

High-Z plasmas have been produced with Ar and/or Ne gas fuelling to increase the ion temperature in Large Helical Device (LHD) plasmas heated with high-energy negative-ion-based neutral beam injection (NBI). Although the electron heating is dominant in the high-energy NBI heating, the direct ion heating power is significantly enhanced in low-density plasmas due to both an increase in the beam absorption (ionization) power and a reduction of the ion density in the high-Z plasmas. Intensive neon- and/or argon-glow discharge cleaning works well to suppress dilution of the high-Z plasmas with wall-absorbed hydrogen. As a result, the ion temperature increases with an increase in the ion heating power normalized by the ion density and reaches 10 keV. An increase in the ion temperature is also observed with the addition of centrally focused electron cyclotron resonance heating to a low-density and high-Z NBI plasma, suggesting improvement of the ion transport. The results obtained in the high-Z plasma experiments with high-energy NBI heating suggest that an increase in the direct ion heating power and improvement of the ion transport are essential to ion temperature rise, and that a high-ion temperature could be obtained as well in hydrogen plasmas with low-energy positive-NBI heating which is planned in the near future in the LHD.

Control of the radial electric field, Er, is considered to be important in helical plasmas, because the radial electric field and its shear are expected to reduce neoclassical and anomalous transport, respectively. In general, the radial electric field can be controlled by changing the collisionality, and positive or negative electric fields have been obtained by decreasing or increasing the electron density, respectively. Although the sign of the radial electric field can be controlled by changing the collisionality, modification of the magnetic field is required to achieve further control of the radial electric field, especially to produce a strong radial electric field shear. In the Large Helical Device (LHD) the radial electric field profiles are shown to be controlled by the modification of the magnetic field by (1) changing the radial profile of the effective helical ripples, ?h, (2) creating a magnetic island with an external perturbation field coil and (3) changing the local island divertor coil current.

Conditions for the excitation of Alfven eigenmodes (AEs) by energetic ions are investigated in neutral-beam-injection (NBI) heated plasmas of the Large Helical Device (LHD). This study is carried out in a wide parameter range of the beta values of the energetic ion components and the ratio of the energetic ion velocity to the Alfven velocity (up to with the assumption of classical slowing down and ). These ranges of parameters cover those predicted for the International Thermonuclear Experimental Reactor (ITER). During this experimental campaign of LHD, toroidicity-induced AEs (TAEs) with n = 1–5 (n being the toroidal mode number), global AEs (GAEs) with n = 0 and 1, and energetic particle modes (EPMs) were observed. The effect of the magnetic configuration on the TAE spectrum was also investigated. In magnetic configurations with relatively high magnetic shear, only TAEs with n = 1 and 2 were observed. On the other hand, TAEs with n up to 5 were observed in magnetic configurations with low magnetic shear. For two typical shots obtained in magnetic configurations characterized by different values of the magnetic shear, eigenfunctions of TAEs were calculated by using a global mode analysis code CAS3D3. The calculated results indicate that the eigenfunctions tend to be localized around the relevant TAE gaps. When the gap is located in the plasma core region (normalized minor radius ρ ≤ 0.4), the TAE tends to become a core-localized type. When the gap is in the outer region (typically 0.5 ≤ ρ ≤ 0.9) of the plasma, the TAE tends to (a) either become a global type having a radially extended structure if the magnetic shear is very weak in the core region inside the gap, (b) or become a gap localized type in the case of finite central magnetic shear. Transition of the eigenmode from the core-localized type with m ~2/n = 1 TAEs (m being the poloidal mode number) to the n = 1 GAEs (or cylindrical AEs) has been observed when the rotational transform at the core ι (0)/2π exceeds the specific value of ι(0)/2π = 0.4.

In order to study erosion/damage processes of the tungsten divertor target, material probe experiments were carried out by exposing tungsten specimens to short pulse helium or hydrogen divertor plasma in LHD (10–20 eV, about 1 × 1022 ions/m2 s). A narrow ‘footprint’, a trace of local-melting, of divertor plasma was formed on the material probe head along the divertor-leg for both hydrogen and helium discharges. In the specimens located rather far from the footprint, where less loading of particles is expected, remarkable blistering was confirmed for both helium and hydrogen irradiation cases. The erosion of the surface was estimated to be 6.3 nm/1 s (one discharge), and 4.1 nm/24 s (19 discharges) for helium and hydrogen, respectively. One should note that erosion due to blistering occurs even for the very low energy hydrogen plasma exposure. It is considered that the abrupt change of specimen temperature due to very high flux particle load may enhance the blistering.

A local island divertor (LID) experiment has begun on LHD, with the aims of controlling edge recycling and improving the plasma confinement. The fundamental divertor functions of the LID have been demonstrated in the recent experiments. From the particle flux profile measurements on the LID head it was found that the particles diffusing out from the core region are well guided along the island separatrix to the LID head. Owing to the closed configuration around the LID head, evidence of the high efficient pumping was observed, together with a strong capacity to screen impurities. The first results of edge modeling using the EMC3-EIRENE code are also presented.

The potential of continuous plasma sustainment of Large Helical Device (LHD) was successfully demonstrated by a one hour discharge with 110 kW electron cyclotron heating (ECH) power. The ECH power of frequency 84 GHz generated by a continuous-work (CW) gyrotron is transmitted through an evacuated waveguide transmission line and injected to the LHD vacuum vessel using waveguide antenna. The plasma density was kept at about 1.5 × 1018/m3 and the electron temperature at the plasma center over 1 keV. Due to the low injection power the density was not so high but the plasma was quite stable. The power injection was terminated manually at 3900 seconds from the limitation of the setting of data acquisition, not for any troubles on devices.

A repetitive pellet injector has been developed for investigation of refueling issues towards the steady-state operation in Large Helical Device (LHD). Continuous operation of more than 10000 pellet launching at 10 Hz has been demonstrated. The maximum repeating rate is 11 Hz. No technical constraint for longer operation has been found. The reliability of pellet launch has exceeded 99.9%. The initial application to the NBI-heated plasmas has been successful in the last experimental campaign of LHD. Although the pulse length is limited by the operational constraint of NBI, the plasma with a density of 8 × 1019 m-3 has been sustained for 2 s by the pellet injection at 10 Hz. A prospect for the future experiment is discussed on the basis of the initial result.

In the Large Helical Device (LHD), energetic ion driven Alfven eigenmodes (AEs) and their impact on energetic ion transport have been studied. The magnetic configuration of the LHD is three-dimensional and has negative magnetic shear over a whole plasma radius in the low beta regime. These features introduce the characteristic structures of the shear Alfven spectrum. In particular, a core-localized type of toroidicity-induced AE (TAE) is most likely because the TAE gap frequency rapidly increases towards the plasma edge. Moreover, helicity-induced AEs (HAEs) can be generated through a toroidal mode coupling as well as poloidal one in the three-dimensional configuration. The following experimental results have been obtained in LHD plasmas heated by tangential neutral beam injection: (1) observation of core-localized TAEs having odd as well as even parity, (2) eigenmode transition of the core-localized TAE to global AEs (GAEs), which phenomenon is very similar to that in a reversed shear tokamak, (3) observation of HAEs of which the frequency is about eight times higher than the TAE gap frequency, (4) enhanced radial transport/loss of energetic ions caused by bursting TAEs in a relatively high beta regime, and (5) seed formation of internal transport barriers induced by TAE-induced energetic ion transport. These results will be important and interesting information for AE physics in toroidal plasmas.

Material probes have been installed at the inner walls along the poloidal direction in large helical device (LHD) from the first experimental campaign. After each campaign, the impurity deposition and the gas retention have been examined to study the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was thoroughly cleaned by glow discharge conditioning and the number of main discharge shots increased. For the 3rd and 4th campaigns, graphite tiles were installed over the entire divertor strike region, and then the wall condition was significantly changed compared with the case of a stainless steel (SS) wall. It was seen that graphite tiles in the divertor were eroded mainly during main discharges, and the SS first wall mainly during glow discharges. During main discharges the eroded carbon was deposited on the entire wall. A reduction of metal impurities in the plasma was observed, which corresponds to the carbonized wall. The deposition thickness was great at the wall far from the plasma. Since the entire wall was carbonized, the amount of discharge gases retained such as H and He became large. In particular, helium retention was large at a position close to the anodes used for helium glow discharge cleanings. One characteristic of the LHD wall is a large retention of helium since the wall temperature is limited to below 368 K. In order to reduce the recycling of the discharge gas, wall heating is necessary.

The 'effective' fuelling efficiency of hydrogen gas puffing ranges from 10% to 50% in the Large Helical Device. A local increase in neutral particle pressure at the gas puff port was measured in the experiment. The pressure increase rate corresponds to ~10% of the gas puff flux. The other 90% of the gas puff flux increases the density and/or the plasma outflow. A particle balance model reveals that the recycling flux estimated from the particle flux on the divertor plates increases during the gas puffing. It is shown that the high effective fuelling efficiency is possibly due to the large recycling flux. At the limit of small recycling flux, the effective fuelling efficiency decreases to ~10%. In the helium gas puff discharge, the effective fuelling efficiency is larger than the hydrogen gas puffing and approaches 100%. This can be related to the large recycling coefficient of more than 0.95.

Standard diagnostics for fundamental plasma parameters and for plasma physics are routinely utilized for daily operation and physics studies in the large helical device (LHD) with high reliability. Diagnostics for steady-state plasma are under intensive development, especially for Te, ne (yttrium–aluminium garnet (YAG) laser Thomson, CO2 laser polarimeter), data acquisition in steady-state and heat-resistant probes. To clarify the plasma properties of the helical structure, two- or three-dimensional diagnostics are being aggressively developed: tangential cameras (fast SX TV, photon counting CCD, Hα TV); tomography (tangential SX CCD, bolometer); imaging (bolometer, ECE, reflectometer). Divertor and edge physics are important key issues for steady-state operation. Diagnostics for neutral flux (Hα array, Zeeman spectroscopy) and ne (fast scanning probe, Li beam probe, pulsed radar reflectometer) are also in advanced stages of development. In addition to these, advanced diagnostics are being intensively developed in LHD through domestic and international collaborations.

Significant progress has been made with ion cyclotron range of frequencies (ICRF) heating in the Large Helical Device. This is mainly due to better confinement of the helically trapped particles and less accumulation of impurities in the region of the plasma core. During the past two years, ICRF heating power has been increased from 1.35 to 2.7 MW. Various wave-mode tests were carried out using minority-ion heating, second-harmonic heating, slow-wave heating and high-density fast-wave heating at the fundamental cyclotron frequency. This fundamental heating mode extended the plasma density range of effective ICRF heating to a value of 1×1020 m−3. This use of the heating mode was its first successful application in large fusion devices. Using the minority-ion mode gave the best performance, and the stored energy reached 240 kJ using ICRF alone. This was obtained for the inward-shifted magnetic axis configuration. The improvement associated with the axis-shift was common for both bulk plasma and highly accelerated particles. For the minority-ion mode, high-energy ions up to 500 keV were observed by concentrating the heating power near the plasma axis. The confinement properties of high-energy particles were studied for different magnetic axis configurations, using the power-modulation technique. It confirmed that with the inward-shifted configuration the confinement of high-energy particles was better than with the normal configuration. By increasing the distance of the plasma to the vessel wall to about 2 cm, the impurity influx was sufficiently reduced to allow sustainment of the plasma with ICRF heating alone for more than 2 min.

Internal transport barriers with respect to electron thermal transport (eITB) were observed in the large helical device, when the electron cyclotron resonance heating (ECH) power was highly localized on the centre of a plasma sustained by neutral beam injection. The eITB is characterized by a high central electron temperature of 6–8 keV with an extremely steep gradient, as high as 55 keV m−1 and a low electron thermal diffusivity within a normalized average radius ρ≈0.3 as well as by the existence of clear thresholds for the ECH power and plasma collisionality.

The behaviour of high-energy ions accelerated by an ion cyclotron range of frequency (ICRF) electric field in the large helical device (LHD) is discussed. A better confinement performance of high-energy ions in the inward-shifted magnetic axis configuration was experimentally verified by measuring their energy spectrum and comparing it with the effective temperature determined by an electron slowing down process. In the standard magnetic axis configuration a saturation of the measured tail temperature was observed as the effective temperature was increased. The ratio between these two quantities is a measure of the quality of transfer efficiency from high-energy ions to a bulk plasma; when this efficiency was compared with Monte Carlo simulations the results agreed fairly well. The ratio of the stored energy of the high-energy ions to that of the bulk plasma was measured using an ICRF heating power modulation method; it was deduced from phase differences between total and bulk plasma stored energies and the modulated ICRF heating power. The measured high energy fraction agreed with that calculated using the injected ICRF heating power, the transfer efficiency determined in the experiment and the confinement scaling of the LHD plasma.

Divertor footprints have been identified within a few mm in accuracy after 10 000 shots. This is the large merit of the large helical device divertor for erosion/deposition studies due to high reproducibility with an external superconducting coils system. Helical distribution of divertor erosion is compared with the prediction from magnetic field characteristics. The measured net erosion depth is found to be about a factor 3 less than the estimated one. Numerical simulations have revealed the net erosion to be very sensitive to deposition of C impurity in the plasma. Eroded carbon atoms are mainly redeposited near the divertor tiles, and partly deposited near the divertor strike point, forming a mixed layer with promptly deposited metals. Deposited metals accumulate locally at the edge of microscale open pores and around grains of graphite. This kind of metals sink possibly plays an important role as an impurity source after the tiles installation. This aspect of ‘microscopic-PSI study’ is very informative for understanding macroscopic-PSI.

Plasma discharges in the Large Helical Device are normally open helical divertor discharges. To compare limiter discharges with open divertor discharges and to examine the role of the peripheral region, a radial movable limiter, whose head was made of carbon with high heat conductivity, was inserted into the plasma from the high field side (near the helical coil). The electron temperature was bounded well by the limiter. A high temperature gradient at the edge region was observed in both open divertor and limiter discharges. Formation of such a high temperature gradient led to good energy confinement even in the limiter discharges and an enhancement factor of 1.1±0.3 for International Stellarator Scaling 95 (ISS95) scaling was observed at every limiter position (0.75

For the third experimental campaign of LHD, graphite tiles were installed in entire region of divertor strike region. Several material probes of SS 316L and graphite were placed along the poloidal direction at #7 toroidal sector. These probes were extracted after the campaign, and impurity deposition on the probe and retention of discharge and impurity gases were analyzed by AES and TDS, respectively. Major impurity species deposited on the probe were C, Fe and O. The iron deposition rate per one main discharge was approximately a half of that in the second campaign. The carbon deposition was significantly large, and the carbon concentration was as high as 90 at.%. These results correspond to the reduction of metal impurity level in the plasma. The retention of discharge gas and impurities doubled compared to the case of the second campaign, due to the carbonized wall. Significant retention of helium was observed, which is one of the characteristics of the LHD wall. The retention of helium is due to charge exchanged He during He main discharge shots and implantation of He ion during He glow discharges.

The first reliable measurement of ion temperature in the divertor plasma of the Large Helical Device has been done by using an ion sensitive probe. The satisfactory current–voltage characteristics of the ion collector for evaluating the ion temperature were obtained at the outer part of the divertor leg. Furthermore, simultaneous ion and electron temperature measurements were successfully done in this part. The results show that the ion temperature is higher than the electron temperature in the part. There is a possibility that the profiles of the evaluated ion temperature which shows relatively higher than the electron temperature at the outside of divertor leg are qualitatively explained by particle’s orbits around the edge and divertor region.

Ion temperature at the plasma centre has been measured from Doppler broadening of Ti XXI (2.61 A) and Ar XVII (3.95 A) x-ray lines using a newly installed crystal spectrometer with CCD detector in ECH, NBI and ICRF plasmas of Large Helical Device (LHD). The ion temperature obtained in a range of 0.6 and 3.5 keV was analysed with electron density and compared with electron temperature. A new parameter range of Ti>Te was found in low-density (ne

Intrinsic impurity behaviour in neutral beam heated LHD long pulse discharges with a discharge duration of 10 s was investigated. Spectroscopic and bolometric measurements showed a remarkable temporal increase of core radiation due to metallic impurities. Central impurity accumulation was found in a narrow plasma density region (e = 1–3×1019 m−3) for both density ramp-up discharges and constant density discharges. In the density ramp-up discharges, high-Z impurities were accumulated in the low-density region and diffused out from the core plasma in the high-density region. The impurity behaviour was compared with neoclassical predictions.

Divertor plasma characteristics in the Large Helical Device (LHD) have been investigated mainly by using Langmuir probes. The three-dimensional structure of the helical divertor, which is naturally produced in the heliotron-type magnetic configuration, is clearly seen in the measured particle and power deposition profiles on the divertor plates. These observations are consistent with the numerical results of field line tracing. The particle flux to the divertor plates increases almost linearly with the line averaged density. The high-recycling regime and divertor detachment, which are observed in tokamaks, have not been observed even during high density discharges with low input power. Both electron density and temperature decrease with increasing radius in the stochastic layer with open field lines, and at the divertor plate they become fairly low compared with those at the last closed flux surface. This means the reduction of pressure along the magnetic field lines occurs in the open field line region in LHD.

Strong nonuniformity in the helical divertor flux distribution is seen both in numerical calculations and in experiments. Flux profiles along the divertor trace on the vacuum vessel change according to vacuum magnetic configurations. Diverted particles choose one of two X-points as the escaping channel to divertor legs, which is determined by the local magnetic shear just inside the X-point. Ergodicity inside both X-points is also estimated quantitatively by a measure of Kolmogorov length for each magnetic configuration. The relationship between ergodicity and the choice of X-point for diverted particles is also discussed.